• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.45-52
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• 2015

Even though a lot of efforts have been devoted to evaluate welding residual stress characteristics of nuclear components, from the view point of accuracy, there are still some arguments in application of engineering estimation schemes. In this paper, three-dimensional finite element analyses (FEA) were carried out to predict residual stress distributions in butt welds of a typical surge line piping. Mesh optimization was conducted and subsequent analysis results such as the axial and hoop stress components along the weld center line and inner wall. Moreover, alternative evaluation was conducted by using three representative equations and their results were compared to those of FEA. Thereby, key parameters affecting to temperature profiles and residual stress distributions were derived as well as an optimum engineering estimation scheme was recommended.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.96-101
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• 2004

Cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal embrittlement at the reactor operating temperature. The objective of this study is to summarize the method of estimating ferrite content, Charpy impact energy and J-R curve and to evaluate the thermal embrittlement of the cast austenitic stainless steel piping used in the domestic nuclear power plants. The result of evaluation, two domestic nuclear power plants used CF-8M and CF-8A material has adequate fracture toughness after saturation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.1-10
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• 2020

This study proposed a simulated test method using ring specimen to evaluate the deformation and failure characteristics of pipe elbows under a large amplitude cyclic load. The validity of the test method was demonstrated by finite element (FE) analysis of pipe elbow and ring specimen under cyclic loads. The results showed that the proposed test method adequately simulates the distribution of circumferential strain at crown of pipe elbows where cracks occur under cyclic loads and presents the cyclic hardening behavior of pipe elbows. The parametric FE analysis showed that consistent simulated test results could be obtained when the test section of the ring specimen is longer than 1/2 of the inner diameter of the ring specimen and the radius of the inner loading jig is less than 1/4 of the inner diameter of the specimen.

• Steel and Composite Structures
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• v.21 no.4
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• pp.703-753
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• 2016

The ratcheting behavior was studied experimentally for Z2CND18.12N elbow piping under cyclic bending and steady internal pressure. Dozens of cyclic plasticity models for structural ratcheting responses simulations were used in the paper. The four models, namely, Bilinear (BKH), Multilinear (MKIN/KINH), Chaboche (CH3), were already available in the ANSYS finite element package. Advanced cyclic plasticity models, such as, modified Chaboche (CH4), Ohno-Wang, modified Ohno-Wang, Abdel Karim-Ohno and modified Abdel Karim-Ohno, were implemented into ANSYS for simulating the experimental responses. Results from the experimental and simulation studies were presented in order to demonstrate the state of structural ratcheting response simulation by these models. None of the models evaluated perform satisfactorily in simulating circumferential strain ratcheting response. Further, improvement in cyclic plasticity modeling and incorporation of material and structural features, like time-dependent, temperature-dependent, non-proportional, dynamic strain aging, residual stresses and anisotropy of materials in the analysis would be essential for advancement of low-cycle fatigue simulations of structures.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.758-766
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• 2018

We developed numerical models to efficiently simulate the low-cycle fatigue behavior of a pipe elbow. To verify the model, in-plane cyclic bending tests of pipe elbow specimens were conducted, and a through crack occurred in the vicinity of the crown. Numerical models based on the erosion method and tie-break method are developed, and the numerical results are compared with experimental results. The calculated results of both models are in good agreement with experimental results, and the model using the tie-break method possesses two times faster calculation speed. Therefore, the numerical model based on the tie-break method would be beneficial to evaluate the strength of piping systems under seismic loadings.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.44-50
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• 2013

Process gas piping is one of the most basic components frequently used in the refinery and petrochemical plants. Many kinds of by-product gas have been used as fuel in the process plants. In some plants, natural gas is additionally introduced and mixed with the byproduct gas for upgrading the fuel. In this case, safety or design margin of the changed piping system of the plant should be re-evaluated based on a proper design code such as ASME or API codes since internal pressure, temperature and gas compositions are different from the original plant design conditions. In this study, series of piping stress analysis were conducted for a process piping used for transporting the mixed gas of the by-product gas and the natural gas from a mixing drum to a knock-out drum in a refinery plant. The analysed piping section had been actually installed in a domestic industry and needed safety audit since the design condition was changed. Pipe locations of the maximum system stress and displacement were determined, which can be candidate inspection and safety monitoring points during the upcoming operation period. For studying the effects of outside air temperature to safety the additional stress analysis were conducted for various temperatures in $0{\sim}30^{\circ}C$. Effects of the friction coefficient between the pipe and support were also investigated showing a proper choice if the friction coefficient is important. The maximum system stresses were occurred mainly at elbow, tee and support locations, which shows the thermal load contributes considerably to the system stress rather than the internal pressure or the gravity loads.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2092-2099
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• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.24 no.5
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• pp.553-560
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• 2015

Fracture mechanics analysis for cracked pipes is essential for applying the leak-before-break (LBB) concept to nuclear piping design. For LBB assessment, crack instability and leak rate should be predicted accurately for through-wall cracked pipes. In a nuclear piping system, elbows are connected with straight pipes by circumferential welding; this weld region is often considered a critical location. Hence, accurate crack assessment is necessary for cracks in the interface between elbows and straight pipes. In this study, the stress intensity factor (SIF) and elastic crack opening displacement (COD) were estimated through detailed 3D elastic finite element (FE) analyses. Based on the results, closed-form solutions of shape factors for calculating the SIFs and elastic CODs were proposed for circumferential through-wall cracks in the abovementioned interfaces under internal pressure. In addition, the effect of the elbow on shape factors was investigated by comparing the results with the existing solutions for a straight pipe.

• Journal of the Korea institute for structural maintenance and inspection
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• v.24 no.3
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• pp.122-129
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• 2020

To conduct probabilistic seismic fragility analysis for nuclear power plants, it is very important to define the failure modes and criteria that can represent actual serious accidents. The seismic design criteria for piping systems, however, cannot fully reflect serious accidents because they are based on plastic collapse and cannot express leakage, which is the actual limit state. Therefore, it is necessary to clearly define the limit state for reliable probabilistic seismic fragility analysis. Therefore, in this study, the limit state of the SCH80 3-inch steel pipe elbow, the vulnerable part of piping systems, was defined as leakage, and the in-plane cyclic loading test was conducted. Moreover, an attempt was made to quantify the failure criteria for the steel pipe elbow using the damage index, which was based on the dissipated energy that used the moment-deformation angle relationship.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.118-123
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• 2004

Cast austenitic stainless steel have been widely used for primary coolant piping in light water reactors. This material is subject to thermal embrittlement at reactor operating temperature. CF-8M and CF-8A cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. Thermal embrittlement results in spinodal decomposition of delta-ferrite leading to decreased fracture toughness. In this study, the specimens were prepared using an accelerated aging method. The measurement of ferrite content, Charpy impact test and J-R test were performed to verify the predicting equation for aged material properties. In case of above 25% ferrite content, predicted result of J-R curve might be non-conservative.